Westinghouse Electric Corporation

Robert D Burack

William C Ritz

Edward Jean Lahoda

David C Grant

William A Byers

John J Wilhelm

Bruce W Bevilacqua

William M Alvino

Wenche W Cheng

Edgar L Kochka

Robert Shaffer

Donald R Stoner

Thomas J Crossley

Fredric W Pement

Thomas J Saska

William J Stenger

Albert J Dietrich

Kenneth E Pfeiffer

Clinton R Wolfe

Rudolph W Brachold

Fredrick J Klorcyzk

Charles E Lutz

Joseph M Gilkison

August J Friedrich

A borated aqueous solution from a nuclear reactor coolant system dilute chemical-decontamination process or from an equipment washing process which has been contaminated with radioactive metals or heavy metals including cobalt, nickel, chromium, iron, manganese, lead or mercury and with a chelating agent, detergent or soap is treated with an oxidizing agent to oxidize the chelating agent, detergent or soap and to precipitate the contaminant metals. The contaminant metals are then separated from the solution by centrifugal filtration, micromembrane belt filtration or magnetic separation. Advantageously, a very small volume of precipitate may then be buried and the decontaminated solution may be recycled or further treated and released for other uses.

Method of treating a contaminated aqueous solution

A nozzle dam assembly for closing off a nozzle of a steam generator and a process for remotely positioning such nozzle dam are disclosed. The nozzle dam includes a mounting ring mounted on an inner surface of the steam generator about a periphery of an inner opening of the nozzle, at least one substantially planar central section and two substantially planar side sections together forming a substantially planar surface, the surface being of a greater diameter than the inner opening of the nozzle. Securing devices are also provided for securing each of the sections to the mounting ring after all the sections are positioned over the inner opening of the nozzle. Also provided is a sealing gasket which is secured to the central section with the sealing gasket having a diameter substantially equal to a diameter of the substantially planar surface formed by the sections for forming a fluid tight seal between a periphery of the substantially planar surface and the inner surface of the steam generator about the periphery of the inner opening of the nozzle. Each of the sections are initially passed through the manway of the steam generator. Once inside the channel head of the nuclear steam generator, the central section of the nozzle dam is retrieved by a robotic arm and positioned over a central portion of the nozzle opening. The robotic arm then positions the remaining side sections over the opening. Once each of the sections are aligned and positioned on the mounting ring, a bolt torque mechanism is then attached to the robotic arm and integral hold down bolts of the sections are sequentially tightened.

Steam generator nozzle dam

In accordance with the present invention, there is described a method and apparatus for concentrating dissolved and solid radioactive materials carried in a waste water solution containing a hazardous chelating agent used for cleaning nuclear equipment. An oxidizing stage for receiving the waste water containing the radioactive materials and the hazardous chelating agent in the presence of an oxidizing agent oxidizes the chelating agent into non-hazardous constituents including gas and water. A separator coupled to the oxidizing chamber receives the waste water containing the radioactive material and separates the radioactive solids from the waste water containing dissolved radioactive materials. An ion exchange chamber containing an ion exchange resin receives the waste water containing the dissolved radioactive materials and removes the same from the waste water by ion exchange with the resin. A dryer receives the radioactive solids from the separator for removing water of hydration therefrom and producing dry solids. A packaging station receives the dry solids and the spent ion exchange resin containing the removed dissolved radioactive materials for packing them in solid form for disposal.

Apparatus for waste disposal of radioactive hazardous waste

Disclosed is a system for driving inspection probes through steam generator tubes of a nuclear power plant, particularly eddy current probes. The system utilizes a single conduit that is connectable to the tube sheet of a steam generator and comprises a multiple-unit probe pusher having a housing containing a plurality of independently actuable drive paths, a multiple-unit probe storage device including a plurality of independently movable take-up reels, and a guide or deflector device for directing any one probe that is driven from any one of the drive paths of the probe pusher to the single conduit. Preferably, the entire system is also remotely controllable. The system enables one probe to be in use for tube inspection while providing readily available spare probes to replace a defective probe with a minimum of downtime and worker radiation exposure.

Multiple-unit probe pusher and system for driving eddy current

Both a system and method for simultaneously heat treating a plurality of heat exchanger tubes in the tubesheet of a steam generator are disclosed herein. The system generally comprises a plurality of heater probes receivable within the open ends of the tubes, each of which has a bladder means for detachably securing the probe within the tube, a plurality of hose-like probe guide assemblies positionable over the open ends of the tubes for guiding the heater probes into the tubes, and an end effector having a cup-shaped coupler connected to a robotic arm for delivering the probe guide assemblies over the open ends of different tubes. The ability of each of the heater probes to detachably mount itself into proper position within its respective heat exchanger tubes advantageously frees up the robotic arm so that it can quickly and sequentially install a number of probe guide assemblies and heater probes in different locations within the tubesheet. Because the probe guide assembly installation time is a relatively small part of the entire heat-treating operation of a single tube, the system and method of the invention allows a plurality of non-adjacent tubes to be substantially simultaneously heat treated, thereby advantageously shortening the total amount of time necessary to perform the heat-treatment maintenance operation.

System and method for stress-relief of welds in heat exchanger tubes

The invention comprises a method for conducting a decontamination solution into a nuclear steam generator for radioactively decontaminating the nuclear steam generator. The invention also comprises controlling the level of the decontamination solution in the heat exchange tubes of the nuclear steam generator in a manner to raise and lower, sequentially, the level of the decontamination solution in the tubes so as to replenish the decontamination solution in the tubes and to carry away the radioactive contaminates.

Flow control method for decontaminating radioactively contaminated

A process for thermally stress-relieving a stressed portion of a metallic tube is disclosed herein. The process generally comprises the steps of inserting a small diameter radiant heater into the open end of the metallic conduit and then positioning it adjacent to the portion to be stress relieved, and heating this portion to a temperature between about 1150.degree. F. to 1500.degree. F. for a time period of between about four and twelve minutes, depending upon wall thickness. In order to determine the proper amount of electrical power to apply to the radiant heater, the emissivity of the stressed portion of the conduit is first measured by heating a portion of the section to incandescence at a known power level, and then determining the temperature of the incandescent tube by means of a pyrometer which compares the intensity of two selected colors or wavelengths of the transmitted light. This process is particularly useful in thermally stress-relieving the U-bends of heat exchanger tubes formed from Inconel.RTM. 600 that are used in nuclear steam engines. It may also be used to stress-relieve such heat exchanger tubes in regions where the reinforcing sleeves have been welded into their interiors. Such thermal stress relieving renders the tubes less susceptible to intragranular stress corrosion cracking.

Process for thermally stress-relieving a tube

A nozzle dam for providing a fluid-tight seal across a primary fluid nozzle in the plenum of a nuclear steam generating vessel includes two fluid-impermeable seals, each comprising a circular three-section foldable seal plate covered with a flexible diaphragm and circumferentially encompassed by an inflatable seal member disposable in frictional sealing engagement with the inner surface of the nozzle. The two seal apparatuses are interconnected by a central tubular coupling coaxial with the nozzle and non-rotatably fixed to each seal, the coupling having a quick-disconnect feature operable from outside the nozzle dam. The coupling provides a conduit for inflation of one of the seals. The two seals are also interconnected by four post assemblies to maintain the seal plates parallel with each other, each post assembly including a cylindrical member pivotally movable between a release position and a supporting position, and provided with a slidable encompassing sleeve to lock it in the supporting position. Structure is provided for pressurizing the inter-seal space and monitoring the pressure in that space.

Dual seal nozzle dam and alignment means therefor

The brazing wand comprises a hollow helically wound electrically conductive tubular member for conducting an electrical current and for conducting a coolant therethrough. The electrical current is used to heat a member to be brazed while the coolant is used to control the temperature of the brazing wand. The brazing wand also comprises an optical sensing device for determining and controlling the brazing temperature.

Brazing wand with fiber optic temperature sensor

The brazing wand comprises a hollow helically wound electrically conductive tubular member for conducting an electrical current and for conducting a coolant therethrough. The electrical current is used to heat a member to be brazed while the coolant is used to control the temperature of the brazing wand. The brazing wand also comprises a flexible cable for conducting the electrical current and coolant to the heating portion of the wand.

Flexible induction brazing wand for hollow tubes

The method comprises applying a braze material to a sleeve in a manner so that the sleeve may be attached to a tube so as to bridge the defective region of the tube. The sleeve is constructed to have a groove on the outside surface thereof that extends around the entire circumference. The groove is filled with braze material and a flux is applied to the braze material in a manner to prevent the braze material and flux from becoming removed. Once such a sleeve has been inserted into the defective tube, the sleeve is internally expanded in the area of the sleeve wherein the groove is located. The expansion of the sleeve in the area wherein the groove is located causes the braze material to closely contact the inside surface of the tube. The other end of the sleeve can then be either similarly placed in contact with the tube or welded to the tube. Next, the area of the tube wherein the braze material is located is internally heated so as to create a brazed joint between the sleeve and the tube. When the joint has cooled, a leaktight joint is formed between the sleeve and the tube.

Sleeve preparation method

A nuclear steam generator modularized wrapper assembly comprising at least two separate generally cylindrical wrapper modules with tube support plates therewithin are assembled in stacked relation and joined within the nuclear containment building, the assembled modules then being placed in the shell with the tube apertures in the tube support plates being aligned with the tube apertures in the tube sheet, and a plurality of adjustable spacer means fixed to the outer circumference of the wrapper assembly are provided and are adjusted to vary the width of the space in the annulus between the wrapper and shell at various locations to lock the assembly in place with axial alignment of all of the apertures in the tube support plates and tube sheets. The method of effecting the replacement of a wrapper assembly in the field within the nuclear containment building includes providing tube aperture keying means at selected locations on the upper surface of the tube sheet to promote alignment of all the tube apertures in the various parts, the keying means being removed after the adjustment of the spacer means in the annulus.

Nuclear steam generator wrapper and shell assembly and method for

The machine comprises a first assembly including a vertical post and a carriage track with a movable carriage thereon, the carriage track being pivotally secured to the upper end portion of the post so that the assembly can be collapsed to a size permitting entry into the channel head through a manway, with the assembly being sufficiently light in weight to permit its installation in the channel head by no more than two individuals, the machine further including a separate horizontal support beam and a separate curved track which are independently separably secured to the vertical post and to the outer end of the carriage track so that a generally quadrantally shaped frame is provided, the carriage track carrying an adjustable decontamination blaster means, and means are provided for moving the carriage along the carriage track and for swinging the carriage track both horizontally and vertically in the frame so that the decontamination means has the capability of sweeping past substantially all of the interiorly facing walls in the channel head. The method of carrying out the decontamination of all of the interiorly facing walls in the space in the channel head is accomplished by traversing the areas to be decontaminated by energizing selective drive means for successive sweeps at rates corresponding to the area to be traversed and then stepping the blaster means to different positions and varying the sweep rate in accordance with the different position.

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Pleasant Hills, PA, United States of America

Robert D Burack